The invention relates to a tube, a bar, a sheet or a strip for a spacer made from zirconium alloy, which has a particularly high resistance to corrosion, both in pressurized water nuclear reactors or PWR and in boiling water reactors or BWR, together with the process for the production thereof. The term tube is understood to mean any tubular article ranging from the blank to the finished tube, e.g. a sheathing or jacketing tube.
French Pat. No. 2 165 270 (1973) describes Zr-Fe-V alloys containing (as % by weight) 0.25 to 1.50% of Fe and 0.1 to 0.6% V and preferably 0.4 to 0.9% Fe and 0.15 to 0.5% V. According to said document, these alloys are able to offer corrosion resistance for several years in pressurized water at 300.degree. C. and can also be used in water-steam mixtures and in superheated steam at 500.degree. C., but for a shorter period.
Such alloys are not used at present and the Applicant has found that those alloys having a Fe+V content exceeding approximately 0.8% had an inadequate cold deformability for conversion into jacketing tubes, structural tubes, bars or sheets.
The Applicant has sought to develop industrial alloys having a very good corrosion resistance in existing nuclear reactors of both the PWR and BWR type.
The following corrosion tests are considered to be representative of existing operating conditions:
PWR: 14 day test at 400.degree. C. in steam under a pressure of 10.3 MPa corresponding to uniform corrosion conditions, the real operating conditions typically giving for the jackets a temperature of 340.degree. to 350.degree. C. in water pressurized at 15 to 16 MPa and at 325.degree. C.
BWR: 24 hour test at 500.degree. C. in steam under 10.3 MPa corresponding to nodular corrosion conditions, the nodules only appearing in this test for jackets having weights higher than approximately 100 mg/dm.sup.2. The real operating conditions typically give for the jackets a temperature of 305.degree. C. in a mixture of water and steam pressurized to 7 MPa and at 285.degree. to 290.degree. C.
In this test, a weight gain of approximately 50 mg/dm.sup.2 is considered to be good, whereas in the preceding test (uniform corrosion), it is common to make use of a zircaloy-4 control.
It is presently desired to increase the service life of jackets and structural parts in the core of PWR and BWR nuclear reactors and if possible to double said life under the existing operating conditions for the same. The problem which the Applicant has attempted to solve corresponds to this desire and at the same time the wish to use a single alloy for both reactor types, whereby the alloy must be easy to cold transform.